Presently, there are approximately 150 metric tons of known weapons grade plutonium and approximately 850 metric tons of known reactor-grade plutonium in the world, with 50 metric tons of reactor-grade plutonium being produced every year. There is likely to be more such plutonium in the world that is unaccounted for. Since these types of plutonium can be used to make weapons of mass destruction, such as thermonuclear bombs and dirty bombs, it is desirable to process any such plutonium so as to render the plutonium difficult to use in making a weapon of mass destruction or to transform any such plutonium into a form that is difficult to use in making any kind of weapon of mass destruction.
Currently, there are two approaches to processing weapons-grade and reactor-grade plutonium such that the end product is either difficult or substantially impossible to use in constructing a weapon of mass destruction. The first approach is to immobilize the plutonium. Typically, this approach involves immobilizing plutonium powder in a glass matrix and then placing the plutonium/glass matrix in a secure storage location. The second approach is to incorporate the plutonium in a nuclear fuel that is burned at a nuclear power plant. The burning of such a fuel results in much of the plutonium being transformed into an isotope that is unsuitable for use in a weapon of mass destruction. Presently, a plutonium-based nuclear fuel that is being used to reduce the supply of plutonium that might be used to produce a weapon is a blend of plutonium-239 and natural or depleted uranium, which is commonly referred to as a mixed oxide fuel (MOX).
There are plutonium-based nuclear fuels suitable for use in a light water reactor (LWR) generating electricity and in which ordinary water is used as a coolant and a moderator to slow down neutrons to the point where their energy ranges fall into the range of higher fission probability. There are two types of LWR, namely, a pressurized water reactor (PWR) and a boiling water reactor (BWR). The plutonium-based nuclear fuel is comprised of plutonium, zirconium hydride, and thorium, which may act as a moderator inside the fuel. In one embodiment, the zirconium hydride comprises 20-50% by weight of the fuel.
Alternatively, the plutonium is less than 10% by weight of the fuel; the zirconium hydride is 20-50% by weight of the fuel; and the thorium is 20-50% by weight of the fuel. Further alternatives of the fuel, have about 40-94% of the plutonium in the fuel as plutonium-239. Other alternative fuel a comprises a zirconium hydride in which the hydrogen to zirconium ratio is in the range of about 1.6-1.8. These fuels may also be used in an LWR reactor, e.g., a TRIGA reactor (Training Research Isotopes General Atomics).
There are benefits to using zirconium hydride alloy fuel in nuclear reactors, at least in part because of its safety characteristics. In fuel the moderator and fuel are intimately mixed. Among the research reactors that commonly use this type is the TRIGA reactor. The NERI program in applying this fuel to the LWR (Z. Shayer and E. Greenspan “Physics Characteristic of U-ZrH1.6Fueled PWR Cores”, PHYSOR 2004, Chicago, Ill., Apr. 25-29, 2004)
The introduction of hydrogen within the fuel permits attainment of neutron moderation to aid plutonium incineration by thermalizing more neutrons, enhancing the neutron absorption probability in the 0.3 eV resonance peak of Pu-239. Use of this fuel may have several advantages over the existing MOX (Mixed Oxide fuel, blends of Uranium and Plutonium oxide) fuel: (a) increased core-life; (b) increased energy generation per fuel loading; (c) reduced waste volume and toxicity due to higher discharge number and to partial utilization of thorium; (d) utilization of thorium resources; (e) improved safety due to the large negative temperature coefficient; (f) improved proliferation resistance by burning up more plutonium and use of thorium; (g) additional significant benefits of the proposed zirconium hydride matrix are better thermal conductivity and fuel storage heat capacity; and (h) the reported experiments with TRIGA fuels indicated low fission gas release.
The neutronic parametric study previously reported is limited directed mainly to infinite pin cell calculations that were performed by WIMSD-5B (WIMSD-5B (98/11), “Deterministic Code System for Reactor-Lattice Calculations”, RSICC CCC-656, user manual (1998), (WIMSD-5B stands for Winfrith Improved Multigroup Scheme Version D-5B, computer code) a deterministic code for reactor core lattice calculations. This code was benchmarked for hydride fuel applications against the well-established codes such as MCNP4B2 and SCALE4.4 codes to provide additional justification of the applicability of the code for the hydride fuel parametric study. Generally there was very good agreement between the codes for various ranges of neutronic parameters and spectrum (Z. Shayer, Neutronic Parametric Analysis: U-ZrH1.6 Unit Cell in PWR NERI Project—Rev.4, NERIO2-189-TM 2 (2003).
The initial analysis shows significant advantages of the proposed fuel over the MOX for incineration of plutonium. Several calculations were performed by the WIMSD-5B to determine the benefit of Pu/ZrHx/Th matrix fuel. FIG. 1 is a sample of the results obtained from this study, which shows the variation of K∞ (Kinf Infinite—Multiplication factor for neutrons)) versus burnup (in GWd/Te ihm; GW days per ton equivalent initial heavy metal [U or Th]) for several fuel types (MOX and oxide fuel). The presence of some Th-232 provides additional fissile material through conversion of Th-232 to U-233, which increases the discharge burnup values to around 80,000 MWd/Te as compared to MOX fuel with LEU (Low Enriched Uranium) that reached only to 65,000 MWd/Te (At K∞=1.03 for a single batch). For a comparison, the discharge burnup value of HEU oxide fuel is only about 45,000 MWd/Te. From FIG. 2 we can see that the destruction rate of Pu-239 is significantly better fast compared to the destruction rate of than the MOX fuel, at 50,000 MWd/Te for MOX fuel, only about 50% of initial Pu-239 is consumed as compared to about 70% for the proposed fuel. This value is increase to 92% for the Pu/ZrHx/Th matrix fuel as compare to only 63% for MOX at 80,000 MWd/Te. The preliminary results show that this fuel is may be suitable ideal for the non-proliferation program to dispose of weapon and power grades plutonium. In this example the calculations were performed for typical PWR rods. The fuel, clad and water temperatures were assumed to be 978 K, 607 K and 579 K, respectively.
Initial results indicated that this Pu/ZrHx/Th matrix fuel would may be very attractive to the disposition of weapon and power grades plutonium. The fuel destruction rates measured in the non-limiting examples described above, were is almost an order of magnitude higher than conventional MOX fuel of containing plutonium. Due to the higher discharge burnup in a smaller core volume, with beneficial safety characteristics and a high prompt reactivity coefficient, a Pu/ZrHx/Th matrix fuel with Zr or SS cladding offers excellent advantages over the conventional MOX fuel for plutonium disposition.